Fuel channel assembly and fuel bundle for a nuclear reactor

ABSTRACT

A fuel assembly for a pressure-tube nuclear reactor includes a fuel channel assembly. The fuel channel assembly has an outer conduit and an inner conduit received within the outer conduit. The conduits define an annular fuel bundle chamber for receiving a flow of a coolant in one direction. The inner conduit includes a central flow passage for receiving a flow of the coolant in an opposite direction. A fuel bundle positioned within the fuel bundle chamber consists of fuel dements arranged to form an inner ring surrounding the inner conduit, and an outer ring surrounding the inner ring. The coolant may be light water, and geometries of the fuel assembly may be selected so moderation by the volume of coolant promotes generally uniform power distribution in the fuel elements.

CROSS REFERENCE TO RELATED APPLICATIONS

This application claims priority to U.S. Provisional Application Nos.61/659,219 and 61/659,229, both filed on Jun. 13, 2012, and U.S.Provisional Application No. 61/731,853 filed on Nov. 30, 2012; theentire contents of each are hereby incorporated herein by reference.

FIELD

The present disclosure relates generally to nuclear reactor technology.

BACKGROUND

The following is not an admission that anything discussed therein isprior art or part of the knowledge of persons skilled in the art.

The Canadian supercritical water cooled reactor (SCWR) is a pressuretube-based reactor concept with heavy water moderator and supercriticallight water coolant. Some features of the Canadian SCWR common toconventional heavy water moderated reactors (HWR) are the separation ofcoolant and moderator by pressure tubes, and the arrangement of fuelpins in annular fuel rings. The Canadian SCWR features common to lightwater (LWR) or boiling water reactors (BWR) are the vertical arrangementof the core, large axial variation in the coolant density andtemperature, and use of long fuel assemblies rather than stacks of shortfuel bundles. The supercritical water based steam cycle in the CanadianSCWR may be similar to that used in existing supercritical-fossil-firedplants.

Safety enhancements in the Canadian SCWR may be achieved through passivesafety features such as a negative power coefficient, negativereactivity coefficients, and/or passive decay heat removal through themoderator. The Canadian SCWR may achieve improvements in both economicsand sustainability through enhanced thermal efficiency, as high as 48%,compared to about 33% for conventional reactors. Improvements ineconomic performance may also be achieved through the simplification ofbalance of plant realized through a direct steam cycle. The use of aplutonium thorium-based fuel cycle in the Canadian SCWR, instead ofenriched uranium, may aid in improved sustainability by reducing theoverall need for mined uranium, thereby extending world uraniumreserves. Enhanced security in the Canadian SCWR may be achieved throughthe use of fuel cycles with increased intrinsic proliferation resistanceand appropriate safeguards.

INTRODUCTION

The following is intended to introduce the reader to the detaileddescription that follows and not to define or limit the claimed subjectmatter.

In an aspect of the present disclosure, a fuel assembly for apressure-tube nuclear reactor may include: a fuel channel assemblyincluding an outer conduit, an inner conduit received within the outerconduit and defining an annular fuel bundle chamber therebetween forreceiving a flow of a coolant in one direction, the inner conduitincluding a central flow passage for receiving a flow of the coolant inan opposite direction; and a fuel bundle positioned within the fuelbundle chamber, the fuel bundle comprising a plurality of fuel elements,and consisting of an inner ring of the fuel elements surrounding theinner conduit, and an outer ring of the fuel elements surrounding theinner ring.

A first ratio of a cross sectional area of the coolant in the fuelbundle chamber and the central flow passage to a cross sectional area ofthe fuel elements may be between approximately 2.6 and 7.5. A secondratio of a cross sectional area of the coolant in the central flowpassage to a cross sectional area of the coolant in the fuel bundlechamber may be between approximately 0.8 and 1.3.

The inner and outer conduits may have generally circular axial crosssectional shapes. The central flow passage may be laterally surroundedby the fuel bundle. A central axis of the central flow passage may belaterally centered relative to the fuel bundle. The fuel bundle may berotationally symmetrical about the central axis. The fuel elements ofthe inner ring may be positioned along a first common circumferenceabout the central axis, and the fuel elements of the outer ring may bepositioned along a second common circumference about the central axisthat is concentric with and laterally outboard of the first commoncircumference. A number of the fuel elements in the inner ring may beequal to a number of the fuel elements in the outer ring. A subchanneldistance between each of the fuel elements in the inner ring and thecorresponding adjacent one of the fuel elements of the outer ring may beapproximately equal to a subchannel distance between each of the fuelelements in the inner ring.

The fuel elements may have generally circular axial cross sections.Axial cross sectional areas of each of the fuel elements in the innerring may be different than axial cross sectional areas of each of thefuel elements in the outer ring. The fuel elements of the inner ring mayhave a smaller cross sectional area than the fuel elements of the outerring.

The fuel channel assembly may include an insulator that is positionedradially intermediate of the fuel bundle chamber and the outer conduit.The insulator may be encapsulated between inner and outer liner tubes,and the outer liner tube may be arranged along an interior surface ofthe outer conduit. The insulator may be formed of a solid material. Theinner and outer liner tubes may be formed of different materials.

A nuclear reactor may include a plurality of the fuel assembliesarranged in a lattice, wherein a moderator region laterally surroundsthe outer conduit of each of the fuel assemblies, the moderator regionretaining a moderator therein. A third ratio of a cross sectional areaof the moderator in the moderator region to a cross sectional area ofthe fuel elements may be between approximately 10 and 20. A forth ratioof a cross sectional area of the moderator in the moderator region to across sectional area of the coolant in the fuel bundle chamber and thecentral flow passage may be between approximately 2.7 and 3.7.

For each fuel assembly, the coolant may flow downwardly in the centralflow passage, and upwardly in the fuel bundle chamber. For each fuelassembly, the inner and outer conduits may be received within a pressuretube. Each pressure tube may include a closed lower end that receivesthe flow of the coolant from the central flow passage, and directs theflow of the coolant into the fuel bundle chamber. The nuclear reactormay further include a first plenum chamber in fluid communication witheach pressure tube to supply the coolant to the central flow passage,and a second plenum chamber in fluid communication with each pressuretube to collect the coolant from the fuel bundle chamber. The coolantmay be light water, and the moderator may be heavy water.

In an aspect of the present disclosure, a fuel assembly for a nuclearreactor may include: a fuel channel assembly including an outer conduit,an inner conduit received within the outer conduit and defining anannular fuel bundle chamber therebetween for receiving a flow of coolantin one direction, the inner conduit including a central flow passage forreceiving a flow of the coolant in an opposite direction; and a fuelbundle positioned within the fuel bundle chamber, the fuel bundleincluding a plurality of fuel elements, wherein at least one of thefollowing conditions is satisfied: (i) a first ratio of a crosssectional area of the coolant in the fuel bundle chamber and the centralflow passage to a cross sectional area of the fuel elements is betweenapproximately 2.6 and 7.5; and (ii) a second ratio of a cross sectionalarea of the coolant in the central flow passage to a cross sectionalarea of the coolant in the fuel bundle chamber is between approximately0.8 and 1.3.

In an aspect of the present disclosure, a nuclear reactor may include: aplurality of fuel assemblies arranged in a lattice, each of the fuelassemblies including a fuel channel assembly including an outer conduit,an inner conduit received within the outer conduit and defining anannular fuel bundle chamber therebetween receiving a flow of a coolantin one direction, the inner conduit including a central flow passagereceiving a flow of the coolant in an opposite direction, and a fuelbundle positioned within the fuel bundle chamber, the fuel bundleincluding a plurality of fuel elements; and a moderator region laterallysurrounding the outer conduit of each of the fuel assemblies, themoderator region retaining a moderator therein, wherein at least one ofthe following conditions is satisfied: (i) a first ratio of a crosssectional area of the moderator in the moderator region to a crosssectional area of the fuel elements is between approximately 10 and 20;and (ii) a second ratio of a cross sectional area of the moderator inthe moderator region to a cross sectional area of the coolant in thefuel bundle chamber and the central flow passage is betweenapproximately 2.7 and 3.7. Both of the conditions (i) and (ii) may besatisfied.

A third ratio of a cross sectional area of the coolant in the fuelbundle chamber and the central flow passage to a cross sectional area ofthe fuel elements may be between approximately 2.6 and 7.5. A forthratio of a cross sectional area of the coolant in the central flowpassage to a cross sectional area of the coolant in the fuel bundlechamber may be between approximately 0.8 and 1.3.

For each fuel assembly, the fuel elements may consist of an inner ringsurrounding the inner conduit, and an outer ring surrounding the innerring. For each fuel assembly, the coolant may flow downwardly in thecentral flow passage, and upwardly in the fuel bundle chamber. For eachfuel assembly, the inner and outer conduits may be received within apressure tube. Each pressure tube may include a closed lower end thatreceives the flow of the coolant from the central flow passage, anddirects the flow of the coolant into the fuel bundle chamber. Thenuclear reactor may further include a first plenum chamber in fluidcommunication with each pressure tube to supply the coolant to thecentral flow passage, and a second plenum chamber in fluid communicationwith each pressure tube to collect the coolant from the fuel bundlechamber. The coolant may be light water, and the moderator may be heavywater.

Other aspects and features of the teachings disclosed herein will becomeapparent, to those ordinarily skilled in the art, upon review of thefollowing description of the specific examples of the presentdisclosure.

BRIEF DESCRIPTION OF THE DRAWINGS

The drawings included herewith are for illustrating various examples ofapparatuses and methods of the present disclosure and are not intendedto limit the scope of what is taught in any way. In the drawings:

FIG. 1A is a schematic cross sectional view of an example of a fuelbundle and a fuel channel assembly;

FIG. 1B is a cutaway perspective view of the fuel bundle and the fuelchannel assembly of FIG. 1A;

FIG. 2A is a perspective view of a pressure-tube nuclear reactorincluding a plurality of fuel channel assemblies;

FIG. 2B is a sectional view of the nuclear reactor of FIG. 2A takenalong line 2 b-2 b;

FIG. 3 is a quarter core channel map and fuel loading scheme for anuclear reactor;

FIG. 4 is a graph of k-infinity modeling data versus axial position inthe nuclear reactor;

FIG. 5 is a graph of CVR modeling data versus axial position;

FIG. 6 is quarter core channel map showing normalized power profiles;

FIG. 7 is a graph of normalized axial power profiles;

FIGS. 8A, 8B, 8C, 8D and 8E are graphs of linear element ratings (LER)for the inner and outer rings of fuel elements;

FIGS. 9 and 10 are graphs of axial power profiles; and

FIGS. 11 and 12 are diagrams of coolant temperature distribution.

DETAILED DESCRIPTION

Various apparatuses or methods will be described below to provide anexample of an embodiment of each claimed invention. No embodimentdescribed below limits any claimed invention and any claimed inventionmay cover apparatuses and methods that differ from those describedbelow. The claimed inventions are not limited to apparatuses and methodshaving all of the features of any one apparatus or method describedbelow, or to features common to multiple or all of the apparatuses ormethods described below. It is possible that an apparatus or methoddescribed below is not an embodiment of any claimed invention. Anyinvention disclosed in an apparatus or method described below that isnot claimed in this document may be the subject matter of anotherprotective instrument, for example, a continuing patent application, andthe applicant(s), inventor(s) and/or owner(s) do not intend to abandon,disclaim or dedicate to the public any such invention by its disclosurein this document.

Recent developments of the Canadian SCWR design include the introductionof a hybrid re-entrant/high efficiency channel. Feeder tubes for coolantflowing out of the fuel channels have been eliminated, and, instead, thefuel channel has a re-entrant or double flow pass configuration. Lightwater coolant flows from an inlet plenum into flow tubes located in thecenter of each fuel channel. The bottom ends of the channels are sealed,and when the coolant reaches the bottom of the central flow tubes, itreaches a space at the bottom of each channel where it is redirectedupward and flows through the region containing the fuel pins orelements.

Various fuel assembly and nuclear reactor arrangements are disclosed inU.S. Provisional Application Nos. 611659,219 and 61/659,229, both filedon Jun. 13, 2012 and entitled A PRESSURE-TUBE NUCLEAR REACTOR WITH A LOWPRESSURE MODERATOR AND FUEL CHANNEL ASSEMBLY.

According to a first aspect of the present disclosure, neutronicsanalysis is used in the design process of the Canadian SCWR is to assessvalues for core physics parameters (e.g., lattice pitch, fissileenrichment of fuel, fuel reload scheme, distribution of burnable neutronabsorbers) that may generally optimize various design targets (e.g.,reactivity coefficients, exit burnup, power peaking factors). Asdescribed in further detail below, the introduction of a central coolanttube, encapsulation of an insulator, and/or reduction in insulatorthickness may impact lattice and core physics performance. Inparticular, these changes may result in a positive increase in coolantvoid reactivity (CVR) and decrease in exit burnup. Additional changes tothe fuel bundle and fuel channel configurations may therefore beintroduced in order to lower the CVR and increase the exit burnup.Relative to previous arrangements, the inner ring of fuel may beremoved, the fuel pin sizes and number of pins may be adjusted toachieve a better power balance between the two remaining rings, and/orthe central flow tube may be expanded. These changes may result in asignificant decrease in CVR and large increase in exit burnup, thusrecovering the target CVR and exit burnup, and may allow a large marginfor additional changes that may be incorporated in the Canadian SCWRdesign.

According to a second aspect of the present disclosure,thermalhydraulics analysis is used to investigate optimization of fuelbundle geometry based on cladding temperatures obtained for axial andradial power profiles corresponding to various design options andconditions. The thermalhydraulic assessment focuses specifically on themaximum wall temperature, which may be an important parameter in thedesign of the fuel bundle. The maximum wall temperature limit may be setbased on the requirement to preserve fuel element integrity, and may berequired to be below 850° C., for example.

Referring to FIGS. 1A and 1B, a fuel bundle 10 is housed in a fuelchannel assembly 12. The fuel channel assembly 12 may be referred toherein as a high-efficiency re-entrant channel (HERC).

In the example illustrated, the fuel channel assembly 12 includes a flowtube or inner conduit 14, which is received within a pressure tube orouter conduit 16. In some examples, the conduits 14, 16 may be receivedwithin one or more additional conduits or pressure tubes (not shown).The inner and outer conduits 14, 16 are illustrated to have generallycircular axial cross sectional shapes. Other cross sectional shapes, forexample but not limited to, square and hexagonal, may be possible. Afuel bundle chamber 18 is defined by an annular space between an outersurface of the inner conduit 14 and an opposed inner surface of theouter conduit 16. The fuel bundle 10 is received in the fuel bundlechamber 18, and laterally surrounds a central flow passage 20.Moderation region 22 laterally surrounds the outer conduit 16.

In some examples, the inner conduit 14 includes the central flow passage20 that accommodates downflow of light water coolant, whereas the fuelbundle chamber 18 receives the corresponding upflow of the coolant.However, in some examples, it may be possible to reverse the orientationof the reactor, so that the plenum is arranged on the bottom of the fuelassembly, and the light water coolant would flow up through the centralflow passage 20 and then down through the fuel bundle chamber 18.Furthermore, arrangements of the fuel assembly other than vertical maybe possible, including, for example, a horizontal arrangement.

The fuel channel assembly 12 may further include an insulator 24. Theinsulator 24 may be sized to be received within the inner surface of theouter conduit 16, so that it is positioned radially intermediate of thefuel bundle chamber 18 and the outer conduit 16. Liner tubes 26 a, 26 bmay encapsulate the insulator 24 to provide a physical barrier betweenthe outer conduit 16 and the insulator 24, and the insulator 24 and thefuel bundle chamber 18, respectively. In some examples, the fuel bundlechamber 18 may include an additional liner tube (not shown) arrangedbetween it and the liner tube 26 b.

In the example illustrated, the fuel bundle 10 includes a plurality offuel elements or pins 28 a, 28 b, which are arranged about the innerconduit 14 into two, generally concentric rings. However, in someexamples, other arrangements of the fuel elements 28 a, 28 b may bepossible, including arrangements having three rings or more, or gridarrangements.

In the example illustrated, the fuel elements 28 a of the inner ring arepositioned along a first common circumference about a central axis 32 ofthe fuel channel assembly 12. The fuel elements 28 b of the outer ringare positioned along a second common circumference about the centralaxis 32, which is concentric with and laterally outboard of the firstcommon circumference. The fuel bundle 10 is rotationally symmetricalaround the central axis 32 of the fuel channel assembly 12, and thenumber of the fuel elements 28 a in the inner ring is equal to thenumber of the fuel elements 28 b in the outer ring (in this case thirtyone each).

In the example illustrated, a subchannel distance between each of thefuel elements 28 a in the inner ring and the corresponding adjacent oneof the fuel elements 28 b of the outer ring may be approximately equalto subchannel distances between each of the fuel elements 28 a in theinner ring. As used herein, “subchannel distance” may refer to theclosest distance between two adjacent fuel elements 28 a, 28 b, whetherit is two adjacent fuel elements 28 a in the inner ring, two adjacentfuel elements 28 b in the outer ring, and/or a fuel element 28 a in theinner ring and the corresponding adjacent one of the fuel elements 28 bof the outer ring. Consistent subchannel geometry may enable a morebalanced heat transfer and coolant mass flow within the fuel bundlechamber 18. Axial cross sectional areas of the fuel elements 28 a of theinner ring may be varied relative to axial cross sectional areas of thefuel elements 28 a of the outer ring to facilitate generally uniformsubchannel geometry. In the example illustrated, the fuel elements 28 a,28 b have generally circular axial cross sectional shapes, and the fuelelements 28 a of the inner ring are smaller than the fuel elements 28 bof the outer ring.

Geometry of the fuel bundle 10 and the fuel channel assembly 12 arrangedin a nuclear reactor may be limited in their performance by unevenradial power distributions. In previous designs for the Canadian SCWR(see, for example, a 78-element fuel assembly having three concentricrows described in J. PENCER, M. EDWARDS, AND N. ONDER, “Axial and RadialGraded Enrichment Options for the Canadian SCWR”, Proc. of the 3rdChina-Canada Joint Workshop on Supercritical Water-Cooled Reactors,CCSC-2012, Van, China, 2012 Apr. 18-20; the entire contents of which arehereby incorporated herein by reference), the fission power may besignificantly higher in the outer ring than inner rings of fuelelements. The uneven power distribution may result in anunderutilization of the inner rings of fuel elements, and may adverselyaffect the fuel performance of the outer ring. In contrast, fuel bundle10 may achieve a nearly even power distribution among the inner andouter fuel rings, thus maximizing fuel utilization while minimizingperformance issues. Secondary benefits may include a reduction (shift inthe negative direction) of CVR, and/or significant increase in latticereactivity and resultant exit burnup.

With implementations of the fuel bundle 10 and the fuel channel assembly12 in a reactor, advantages of balanced radial power, lower CVR andhigher reactivity may be achieved through a balance of neutronmoderation by the heavy water in the moderation region 22, absorptionand fission in the fuel elements 28 a, 28 b, moderation and absorptionin the light water coolant in the central flow passage 20, andmoderation and absorption in the light water coolant surrounding thefuel elements 28 a, 28 b within the fuel bundle chamber 18. Moderationin the heavy water in the moderation region 22 may drive fission in thefuel elements 28 b of the outer ring, while moderation in the lightwater coolant in the central flow passage 20 may drive fission in thefuel elements 28 a of the inner ring. Balance between these twocontributions to the lattice physics behavior may be characterized bythe following lattice parameters: (i) ratio of the total cross sectionalarea of the coolant (in the fuel bundle chamber 18 and the central flowpassage 20) to the cross sectional area of the fuel elements 28 a, 28 b;(ii) ratio of the cross sectional area of the coolant in the centralflow passage 20 to the cross sectional area of the coolant surroundingthe fuel bundle 10 in the fuel bundle chamber 18; (iii) ratio of themoderator cross sectional area to the cross sectional area of the fuelelements 28 a, 28 b (in one lattice cell); and (iv) ratio of the crosssectional area of the moderator to the total cross sectional area of thecoolant (in one lattice cell).

Lattice level scoping studies were performed to examine the impact ofvariation of the parameters above on the ratio of power densities of theouter to inner fuel rings, lattice CVR, and infinite lattice neutronmultiplication factor, k-infinity (ratio of neutrons produced toneutrons absorbed). The target for the ratio of outer to inner fuelpower densities is 1, the lattice CVR target is to be negative, and thek-infinity target is to be maximized. Based on the lattice scopingstudies, the following ranges of parameters (applied simultaneously)were found to yield values for power density ratio, CVR and k-infinitythat satisfied the targets for lattice physics performance: acoolant-to-fuel ratio of between approximately 2.6 and 7.5; aninner-to-outer coolant ratio of between approximately 0.8 and 1.3; amoderator-to-fuel ratio ranging of between approximately 10 and 20; anda moderator-to-coolant ratio of between approximately 2.7 and 3.7. Basedon the relatively wide ranges in the moderator-to-fuel andcoolant-to-fuel ratios, the total fuel mass in the fuel assembly may bevaried significantly without adverse impacts to power density ratio, CVRor k-infinity. However, other design constraints may restrict thevariation in fuel mass in the assembly, such as the maximum allowablepower density.

Referring now to FIGS. 2A and 2B, an example of a pressure-tube nuclearreactor 100 includes an outer shell vessel 102, also referred to as acalandria, which is configured to contain a moderator fluid, e.g., heavywater, The calandria 102 includes a bottom wall 104 and a side wall 106extending upward from the bottom wall 104 about a calandria axis 108,and terminating at an upper rim 110. In the example illustrated, thecalandria 102 is shown as being generally circular in axialcross-sectional shape.

Referring to FIG. 2A, a pressurized coolant plenum vessel 112 isconnected to the upper rim 110 of the calandria 102. The plenum vessel112 is configured to supply coolant to a plurality of the fuel channelassemblies 12 (FIG. 2B) in the reactor 100 and to extract the heatedcoolant from the fuel channel assemblies 12 after the coolant has beenheated by flowing past the fuel bundles 10 (FIGS. 1A and 1B) containedwithin the fuel channel assemblies 12. The coolant is pressurized to ahigher pressure than the moderator, and the plenum vessel 112 is apressure vessel capable of withstanding the operating temperatures andpressures of the coolant.

Referring to FIG. 2B, in the illustrated example the plenum vessel 112includes a tubesheet 114, a sidewall 116 and a cover 118 that cooperateto define a first plenum chamber 120. In the example illustrated, thefirst plenum chamber 120 is in fluid communication with each of the fuelchannel assemblies 12 to allow coolant to flow between the first plenumchamber 120 and each of the fuel channel assemblies 12. Only a singlefuel channel assembly 12 is illustrated for clarity. One or more fluidports may be provided in the plenum vessel 112 to allow coolant to flowin and out of the first plenum chamber 120. In the example illustrated,the plenum vessel includes four ports 122 spaced apart from each otheraround the sidewall 116.

In the example illustrated, a second plenum 124 is nested within thefirst plenum chamber 120. The second plenum 124 includes a second plenumchamber 126 that is bounded by a bottom wall 128, a sidewall 130 and anupper wall or lid 132. In this configuration, the second plenum chamber126 is self-contained and is fluidly isolated from the first plenumchamber 120 so that fluid within the second plenum 124 does not mix withfluid in the first plenum chamber 120.

in the example illustrated, the second plenum 124 is sized such that awidth 140 of the second plenum 124 is less than an internal width 142 ofthe first plenum chamber 120. In this configuration, a gap around theperimeter of the second plenum 124, between an outer surface of thesecond plenum sidewall 130 and an opposing inner surface of the firstplenum sidewall 116, provides a fluid flow path around the outside ofthe second plenum 124 to link the upper portion 136 and lower portion138.

With continued reference to FIG. 2B, the bottom wall 128 of the secondplenum 124 includes a plurality of apertures for connecting to each ofthe fuel channel assemblies 12 to allow coolant fluid to flow betweeneach of the fuel channel assemblies 12 and the second plenum chamber126. In this configuration, a coolant flow path is provided so thatcoolant fluid may flow between the first plenum chamber 120 and thesecond plenum chamber 126 via the fuel channel assemblies 12.Optionally, some or all of the fuel channel assemblies 12 may bedetachably coupled to the bottom wall 128 using any suitable connector.Providing detachable connections may allow individual fuel channelassemblies 12 to be removed, serviced and/or replaced without requiringreplacement of other fuel channel assemblies 12.

In the example illustrated, the first plenum chamber 120 functions as acoolant inlet plenum, the ports 122 operate as coolant inlet ports, thesecond plenum chamber 126 functions as a coolant outlet plenum and itsports 134 (FIGS. 2A) function as coolant outlet ports. In thisconfiguration, the first plenum chamber 120 is in fluid communicationwith each fuel channel assembly 12 to supply coolant and the secondplenum chamber 126 is in fluid communication with each fuel channelassembly 12 to collect the heated coolant fluid. The coolant flows fromthe first plenum chamber 120 in a downward direction 144 through thecentral flow passage 20 (FIGS. 1A and 1B) of the fuel channel assemblies12, and flows to the second plenum chamber 126 in an upward direction146 through the fuel bundle chambers 18 (FIGS. 1A and 1B) of the fuelchannel assemblies 12. The fuel bundles 10 are positioned within thefuel channel assembly 12 in the flow path of the coolant flowing in theupward direction 146.

Coolant may be supplied to the first plenum chamber 120 at any suitableinlet temperature and inlet pressure. The combination of inlettemperature and pressure desired may be based on the properties of agiven reactor core design and/or nuclear fuel type. Optionally, theplenum vessel 112 may be configured to handle liquid coolants, gascoolants, mixed-phase coolants and supercritical coolant conditions. Forexample, in some configurations the inlet temperature may be betweenabout 100° C. and about 370° C. or more, and may be between about 260°C. and about 350° C. The inlet pressure may be between about 5 MPa andabout 30 MPa or more, and may be between about 10 MPa and about 26 MPa.Optionally, the inlet conditions may be selected so that the incomingcoolant remains subcritical. This may help facilitate greater energypickup from the fuel bundles 10 when the coolant flows through the fuelchannel assemblies 12.

The temperature and pressure of the coolant as it exits the fuel channelassemblies 12, and flows into the second plenum 124, may vary based onthe quantity of heat transferred from the fuel bundles 10 and the amountof pressure drop generated as the coolant flows past the fuel bundles 10in the fuel channel assembly 12. When operated in subcriticalconditions, the outlet temperature may be between about 290° C. andabout 350° C. and the outlet pressure may be between about 8-12 MPa.When operated under supercritical coolant conditions, the outlettemperature may be between about 374° C. and about 675° C. (and may beabout 625° C.) and the outlet pressure may be about 22-26 MPa. In otherconfigurations, the outlet pressure may be between about 12 and about 22MPa and may be greater than 26 MPa.

In some configurations, there may be a significant temperaturedifference between the coolant in the first plenum chamber 120 andcoolant in the second plenum 124. For example, under supercriticaloperating conditions, the temperature difference between the inletplenum 120 and outlet plenum 124 may be between about 250-300° C., ormore. Such temperature differences may impart significant thermalstresses in the lid 132, sidewall 130 and bottom wall 128. As it islocated outside of the neutron field, the outlet plenum 124 may be madefrom a variety of suitable materials, including, for example stainlesssteel and nickel-based super alloys. Optionally, some or all of theoutlet plenum 124 may be thermally insulated using any suitabletechniques and/or materials to help limit heat transfer between theplenums 120, 124 and to help reduce thermal stresses on the secondplenum 124. For example, the lid 132, sidewall 130 and bottom wall 128may be coated with an insulating material and/or made from multiplelayers. Alternatively, as the pressure difference between the plenumsmay be relatively small, the second plenum 124 may be formed frommaterials that have desirable thermal properties, including, for examplerefractory materials and ceramic-based materials, instead of highlythermally conductive metals.

Heated coolant extracted from the second plenum 124 will often be usedto generate electrical power. Optionally, the heated coolant fluid maybe used to directly drive suitable steam turbine generators (not shown).This may help improve the efficiency of a nuclear power generationstation as the heated coolant may remain at a high temperature when itreaches the turbines. Alternatively, the heated coolant may be used toheat a secondary circuit, for example via a steam generator, and theturbine generators may be driven by steam in the secondary circuit.Configuring the system to include a steam generator and secondarycircuit may help increase the safety of the power generation system, butmay reduce overall efficiency.

In the example illustrated, the tubesheet 114, in combination with thefuel channel assemblies 12, forms part of the pressure barrier betweenthe high pressure coolant and the low pressure moderator. The tubesheet114 may also separate the reactor core (containing fissile nuclear fuel)from the non-core portions of the reactor. The tubesheet 114 includes aplurality of apertures 148 to accommodate the plurality of fuel channelassemblies 12.

Referring to FIG. 2B, a plurality of fuel channel assemblies 12 arearranged in a lattice 150, and each extends from the plenum vessel 112into the calandria 102. Portions of each of the fuel channel assemblies12 are submerged in the moderation region 22. The number, configurationand arrangement or pitch spacing of the apertures 148 in the tubesheet114 (defined as generally horizontal distance 152 between fuel channelaxes 154 within the lattice 150) may be any suitable distance.

In the example illustrated, fuel channel assembly 12 includes a pressuretube 158, and the inner and outer conduits 14, 16 (FIGS. 1A and 1B) arereceived within the pressure tube 158. The pressure tube 158 of each ofthe fuel channel assemblies may be sealed to the tubesheet 114, andprovide both pressure and fluid separation between the moderator in themoderation region 22 and the coolant circulating within the fuel channelassemblies 12.

In the example illustrated, an upper end 160 of each of the fuel channelassemblies 12 is connected to the bottom wall 128 of the second plenum124. Adjacent to the bottom wall 128, and above the tubesheet 114, inletports and feeder conduits provide fluid communication between the firstplenum chamber 120 and the central flow passage 20 (FIGS. 1A and 1B).The inlet ports are arranged about the outer conduit 16, and the feederconduits provide a fluid-sealed connection between the inlet ports andthe central flow passage 20, while permitting flow of the heated coolantupwardly through the fuel bundle chamber 18.

Each of the fuel channel assemblies 12 extends downwardly generallyparallel to the axis 108 from the upper end 160 to a closed lower end162. The length of each fuel channel assembly 12 may be selected to beany suitable length that is compatible with other components of thereactor 100, and may be, for example, between about 1 m and about 10 m.The axial length of the fuel bundle 10 within each fuel channel assembly12 may be any suitable length, and may be between about 0.2 m and about5 m or more.

In the example illustrated, the fuel bundle 10 is made up of a singlebundle of the fuel elements 28 a, 28 b. In other examples, the fuelbundle may be formed of multiple portions arranged axially along thelength of the fuel assembly 12 within the moderation region 22. Multipleportions facilitate batch-fueling of the reactor 100 axially, as well asradially within the lattice 150.

The closed lower end 162 contains the coolant exiting the central flowpassage 20 (FIGS. 1A and 1B), and directs the coolant upward into thefuel bundle chamber 18. The inner conduit 14 may optionally extendaxially beyond a lower face of the fuel bundle 10 in the fuel channelassembly 12 by an extension length. Extending the inner conduit 14beyond the lower face of the fuel bundle 10 may help balance thedistribution of coolant flow within the fuel bundle chamber 18 before itreaches the lower face. Optionally, one or more flow directors (notshown), including for example a baffle, vane, guide or other flowdirecting apparatus, may be provided axially between the lower face ofthe fuel bundle 10 and a lower end of the inner conduit 14 to helpmodify or balance the coolant flow as it enters the fuel bundle chamber18. Extending the inner conduit 14 beyond the lower face may also helpaccommodate thermal expansion (lengthening) and/or creep of the fuelbundle 10 when the reactor 100 is in use, to help provide coolant flowto the lowermost portion of the fuel bundle 10.

While the present disclosure refers to the use of the fuel bundle 10 andfuel channel assembly 12 specifically in the context of the CanadianSCWR design, it should be appreciated that the teachings herein may beapplicable to other nuclear reactor designs.

Reference is now made to the following description of neutronics andthermalhydraulic analysis of exemplary configurations of the fuel bundle10 and fuel channel assembly 12, which is intended to be illustrativebut non-limiting.

1. Neutronics Analysis

A. Geometrical and Material Specifications

The outermost component of the HERC is the outer conduit 16, which maybe formed of an Excel (zirconium-based) alloy. In some conventionalpressure tube reactor fuel channel designs, the pressure tube isseparated from the moderator by a CO₂ filled gap surrounded by acalandria tube. However, for the HERC design, the calandria tube and gapare omitted and the pressure tube 16 may be in direct contact with theheavy water in the moderation region 22. The insulator 24 may be solid,encapsulated zirconia, and may be located directly inside the outerconduit 16, and isolate it from the high temperatures in the coolant.The insulator 24 may be supported on its outer surface by the liner tube26 a, which may be formed of a solid Excel alloy. The insulator 24 maybe supported on its inner surface by the liner tube 26 b, which may beformed of solid zirconium-modified stainless steel. The coolant entersat a top (not shown) of the inner conduit 14 into the central flowpassage 20. The coolant then flows down the central flow passage 20,reaches the bottom (not shown) and is directed upwards into the fuelbundle chamber 18, where it is heated up by the fuel bundle 10.

Exemplary specifications for the fuel bundle 10 and the fuel channelassembly 12 are given in Table 1.

TABLE 1 Fuel bundle and fuel channel assembly exemplary specifications.Density Component Dimension Material Composition (wt %) (g/cm³) CentralFlow 4.45 cm radius Light Water 100% H₂O 0.59254 Passage 20 InnerConduit 4.45 cm inner Zr-modified C: 0.034; Si: 0.51; 7.90 14 radius(IR) 310 Stainless Mn: 0.74; P: 0.016; 0.1 cm thick Steel S: 0.0020; Ni:20.82; (Zr-mod SS) Cr: 25.04; Fe: 51.738; Mo: 0.51; Zr: 0.59 Fuel 0.415cm 15 wt % Pu: 13.23; Th: 74.70; 9.91 Elements 28a radius PuO₂/ThO₂ O:12.07 (inner ring) 5.30 cm pitch circle radius no displacement angleFuel 0.465 cm 12 wt % Pu: 10.59; Th: 77.34; 9.87 Elements 28b radiusPuO₂/ThO₂ O: 12.08 (outer ring) 6.55 cm pitch circle radius nodisplacement angle Fuel Cladding 0.06 cm thick Zr-mod SS As above 7.90Coolant n/a Light Water 100% H₂O Variable Liner Tube 7.20 cm IR Zr-modSS As above 7.90 26a 0.05 cm thick Insulator 24 7.25 cm IR Zirconia Zr:66.63; Y: 7.87; O: 25.5 5.83 0.55 cm thick (ZrO₂) Liner Tube 7.80 cm IRExcel Sn: 3.5; Mo: 0.8; Nb: 0.8; 6.52 26b 0.05 cm thick (Zirconium Zr:94.6 Alloy) Outer Conduit 7.85 cm IR Excel Sn: 3.5; Mo: 0.8: Nb: 0.86.52 16 1.2 cm thick (Zirconium Zr: 94.9 Alloy) Moderation 25 cm squareD₂O 99.833% D₂O; 0.167% 1.0851 region 22 lattice pitch H₂O na na Rg-PuPu-238:2.75; Pu- 239:51.96; Pu-240:22.96; Pu- 241:15.23; Pu-242:7.10

The inner conduit 14 may be a solid tube of zirconium-modified stainlesssteel. The inner conduit 14 prevents mixing of the downward flowingcoolant with the upward flowing coolant. In some examples, although notshown, the inner conduit 14 may include an insulating layer in order toprevent heat transfer to the downward flowing coolant.

In the example illustrated, the fuel bundle 10 has two concentric fuelrings, each with 31 elements, which may be composed of mixtures ofthorium dioxide and plutonium dioxide. However, in some examples, otherconventional combinations of nuclear fuel may be used. For example, theconcepts described herein may also work with enriched urania fuel, andmay work with an enriched urania and thoria mixture-based fuel. The fuelelements 28 a, 28 b may be clad in 0.6 mm thick zirconium-modified 310stainless steel.

For the purposes of calculations herein, the PuO₂ may be the oxide formof reactor grade plutonium (Rg-Pu), which may be recycled from usedlight water reactor fuel. The Pu isotopic composition is based onprevious samples. The thorium may be assumed to be isotopically pureTh-232. The theoretical densities of pure PuO₂ and ThO₂ used herein were10.0 g/cm³ and 11.5 g/cm³, respectively, based on data in previousstudies. For the (Pu—Th)O₂ mixtures, it was assumed that the densitiesof the mixtures were simply the volume weighted averages of thecomponents. It was further assumed that the fuels in pellet form haddensities equal to 97% times the theoretical density.

Grid spacers 30 (FIG. 1B) and/or wire wrap (not shown) may be includedin the fuel bundle 10 and the fuel channel assembly 12, and appendagesmay also be added to promote turbulence within the coolant flow. Theparasitic absorption in these additional materials may be accounted forin the lattice modeling via an increase in the density, or “smearing”,of the stainless materials already present in the modeling herein. Thisdensity adjustment is not performed in the current models.

B. Core Geometry and Refueling Scheme

The reactor core is batch-fueled using three batches arranged radially.It is designed to generate 2540 MW of thermal power corresponding to1200 MW of electric power assuming a 48% thermodynamic cycle efficiency.The core consists of 336 fuel channel assemblies, each containing a 500cm long fuel bundle arranged in a 25 cm square pitch lattice. The corediameter is 625 cm and core height is 600 cm including 50 cm thick lowerand upper axial D₂O reflectors. The channel layout and refueling schemeare shown schematically in FIG. 3. No neutron absorbers or reactivitydevices have been incorporated in the core.

C. Impact on Lattice Physics

As discussed above, a number of changes have been made to the fuelbundle and fuel channel assembly design versus previous versions (i.e.the 78-element fuel assembly having three concentric rows). Thesechanges are: addition of the inner conduit 14 with the central flowpassage 20 for re-entrant coolant flow; encapsulation of the insulator24 by the liner tubes 26 a, 26 b, isolating it from the coolant in thefuel bundle chamber 18; and/or reduction of the thickness of theinsulator 24 and change from a porous to solid material.

The various changes above were investigated for fresh fuel vialattice-based calculations of k-infinity. A software package was used inconjunction with a nuclear data library. The changes to the fuel bundleand channel options were applied cumulatively, starting with the78-element fuel assembly (described in J. PENCER, M. EDWARDS, AND N.ONDER, “Axial and Radial Graded Enrichment Options for the CanadianSCWR”, Proc. of the 3rd China-Canada Joint Workshop on SupercriticalWater-Cooled Reactors, CCSC-2012, Xi'an, China, 2012 Apr. 18-20) as areference, and compared to the present 62-element design. Plots ofk-infinity as a function of axial position along the channel are shownin FIG. 4, and corresponding plots of infinite lattice CVR are shown inFIG. 5.

As described herein, changes in reactivity and CVR may be a result ofneutron moderation in the coolant region, particularly in the centralflow passage 20. The overall impact of the coolant region on excessreactivity (or k-infinity) and CVR may be understood in the context ofthe impact of lattice pitch and fuel enrichment on SCWR latticereactivity and CVR. It has been shown that it is possible to shift theCVR in the negative direction via reduction in the lattice pitch (LP).It has also been observed that the dependence of CVR on LP may bedetermined by the balance between moderation and absorption of neutronsin the coolant. For the cooled lattice, as lattice pitch is decreased,the positive reactivity contribution from neutron moderation in thecoolant may dominate over the negative reactivity contribution fromneutron absorption. Thus, a lattice pitch may be selected such that thereactivity contribution from neutron moderation in the coolant maydominate over the contribution from absorption. On coolant voiding, thenegative reactivity contribution due to the decrease in neutronmoderation may therefore dominate over the positive reactivitycontribution due to the decrease in neutron absorption, leading to a netdecrease in lattice reactivity, and hence a negative coolant voidreactivity.

Coolant in the central flow passage 20 (FIG. 1A) may increase k-infinityand may shift the CVR in the negative direction. Both of these resultsmay be a consequence of the additional volume of coolant within the fuelchannel assembly 12; the coolant in the central flow passage 20 acts asa moderator, thus increasing the reactivity under normal conditions andleading to a negative contribution to CVR when the moderation is lost.The CVR may therefore be “tuned” or shifted by varying the volume ofcoolant in the central flow passage 20, which may be achieved bychanging the inner diameter of the central flow passage 20. Generally,an increase in the flow tube inner diameter will shift the CVR in thenegative direction, while a decrease in the flow tube inner diameterwill shift the CVR in the positive direction.

Encapsulation of the insulator 24 may decrease k-infinity and shift CVRin the positive direction. Without encapsulation, for example, a 76% (byvolume) porosity insulator may contain a significant (approximately 25%)amount of the coolant inside the fuel channel. Thus, encapsulation ofthe insulator 24, which eliminates this volume of coolant water, mayresult in a significant loss of neutron moderation in the coolantregion, leading to both a reduction in k-infinity and positive change inCVR. The change from a perforated to solid stainless steel liner tubemay increase the amount of stainless steel between the fuel bundle 10and the moderation region 22, increasing the parasitic neutronabsorption, which also may contribute to the reduction in k-infinity.

Reduction of the insulator thickness and change from a porous to a solidinsulator may increase the amount of coolant inside the fuel channelassembly 12, but also increases the amount of material between the fuelbundle 10 and the moderation region 22. The net result may be a decreasein k-infinity, due to an increase in the parasitic absorption in thesolid insulator, and decrease in CVR, due to the increased moderation inthe coolant.

In summary, in comparison with the 78-element fuel assembly according toa previous design, removal of the innermost ring of fuel elements andexpansion of the inner conduit 14 may result in an increase in theoverall reactivity and shift the CVR in the negative direction. Both ofthese results may be a consequence of increased moderation in thecoolant in the central flow passage 20. The combination of the twoconcentric rings of fuel elements 28 a, 28 b in the fuel bundle 10, andsignificant moderation in the central flow passage 20 of the fuelchannel assembly 12, may also aid in balancing the radial powerdistribution within the fuel assembly.

D. Core Physics

Core calculations were performed using a software package with a nucleardata library based on the specifications described above. Values forintegral core parameters (e.g., average exit burnup, etc.) are listed inTable 2, along with corresponding values obtained previously based onthe reference 78-element fuel assembly.

TABLE 2 Comparison of integral core parameters. Parameter 78-Element62-Element Average initial wt % PuO₂  13%  13% Average initial fissilewt % heavy element 8.7% 8.7% Average Exit Burnup (MWd/kg) 41.5  58.6 Cycle Length (EFPD) 455    425    Excess Reactivity BOC/EOC (mk)95.3/9.7  108.9/10.0  Coolant Void Reactivity BOC/EOC (mk) −4.4/−5.7−30.4/−45.2 Fuel Temperature Coefficient BOC/EOC nd/nd −0.05/−0.05(mk/K) Moderator Temperature Coefficient nd/nd −0.12/−0.11 BOC/EOC(mk/K) Channel Power Peaking Factor BOC/EOC 1.28/1.19 1.31/1.22 AxialPower Peaking Factor BOC/EOC 1.39/1.19 1.18/1.05 Maximum LER (kW/m)37.4  41.3  Exit [fissile Pu] (wt % HM) 4.5 2.7 Exit [U-233 + Pa-233](wt % HM) 1.1 1.1

The fuel bundle 10 and the fuel channel assembly 12 in accordance withthe present design show some performance enhancements over the previousdesign. There is an almost 50% increase in exit burnup. The extendedburnup of the present design also results in a significantly lower (⅓less) remainder of fissile Pu at the end of the cycle. There is a 15%decrease in the beginning of cycle (BOC) axial peaking factor, and asimilar decrease at the end of cycle (EOC). These gains are slightlyoffset by the reduction in cycle length by about 5% (which reduces thecapacity factor), and increase in radial power peaking factor by about2%. There is also a significant decrease in the core average CVR, whichis discussed in more detail below. The fuel temperature (FTC) andmoderator temperature (MTC) coefficients were not evaluated previously.For the present design, the FTC and MTC are negative at BOC, EOC andthroughout the cycle.

The differences between the two designs may be mainly due to theintroduction of the inner conduit 14 with the central flow passage 20.As discussed above, the coolant in the central flow passage 20 mayprovide a significant amount of neutron moderation. This increasedmoderation may lead to an increase in net reactivity, which may resultin an increase in the maximum achievable exit burnup. The moderation inthe central flow passage 20 does not generally change with axialposition, and so the axial power profile may vary much less than in theprevious design, thus reducing the axial power peaking factor. Thereduction in cycle length may be a result of the reduction in fuel massrelated to the change in the design, but this reduction may be nearlyoffset by the increase in initial reactivity. The increase in channelpower peaking factor may be due to the larger initial reactivity offresh fuel, and resultant increase in reactivity difference betweenfresh and partially irradiated fuel.

The CVR of the core of the present design may be negative and itsmagnitude may be quite large (e.g., ranging from −30 mk to −45 mk). Anegative CVR may be desirable because of the safety advantage ofnegative reactivity feedback. As discussed above, the CVR may be variedby changing the inner diameter of the central flow passage 20 of theinner conduit 14 An appropriate range for the core average CVR maytherefore be achieved through selection of flow volume in the centralflow passage 20.

Referring to FIG. 6, a quarter core channel map of normalized channelpower profile is provided at the beginning of cycle (BOC) and the end ofcycle (EOC). This power profile is similar to that observed previouslywith the 78-element fuel assembly, although the peak channel powers areslightly higher. The highest channel powers correspond to the fuelchannel assemblies having fresh fuel, and it is anticipated that somedegree of power leveling may be achieved with the addition of burnableneutron absorber in the fresh fuel, or variation in fresh fuelenrichment.

The normalized axial power profiles for BOC and EOC for the peak powerchannel are plotted in FIG. 7. The power profile at BOC is slightlyasymmetric, with a maximum located approximately 3 m from the bottom ofthe fuel assembly. At EOC, the axial power profile is symmetric, withmaxima located at approximately 0.5 m above the bottom of the fuelassembly and 0.5 m below the top of the fuel assembly. Comparison of theBOC and ECG axial profiles shows that there is a flattening of the powerprofile over time. This power flattening may be a result of thecompensatory effect of reactivity on neutron leakage. At BOC, thedistribution of fissile material is axially uniform. The power is highertoward the center of the fuel channel because of neutron leakage at thetop and bottom of the channel. The higher power in the center of thechannel at BOC may result in a higher rate of depletion of fissilematerial in this region and simultaneous higher production of neutronabsorbing fission products. The depletion of fissile material in thecenter of the channel and buildup of fission products then may lead to adecrease in power toward the center of the channel relative to the ends,resulting in a flattening of the power profile, which is seen at EOC.

Previous results obtained for the 78-element fuel assembly are alsoplotted in FIG. 7. The present design with 62-elements shows asignificant reduction in axial power peaking compared with the previousdesign at both the BOC and EOC. In addition, the asymmetry in the axialpower shape previously observed with the 78-element fuel assemblyappears to be absent for the present design. The axial power profilesfor the present design therefore show significant improvement over theprevious design.

The power history of the fuel assembly in the peak power channel wasextracted from a calculation, and used in subsequent calculations toobtain the fuel element power distribution as a function of time andaxial location along the fuel channel assembly. The resultant linearelement ratings (LER) for the fuel elements in the inner and outer ringsare plotted as a function of time and at various axial locations inFIGS. 8A, 8B, 8C, 8D and 8E (the distances refer to distances from thebottom of the fuel assembly.). The abrupt changes in LER at 425equivalent full power days (EFPD) and 850 EFPD correspond to the changein channel power when the fuel assembly is moved to a new channelposition during refueling.

The maximum LER is approximately 40 kW/m, and occurs in the outer fuelring, at the BOC, near the center of the fuel channel. The maximum LERoccurs at the same time and location as the highest reactivity andhighest channel power during the cycle. The largest difference in LERbetween the inner and outer fuel rings occurs at the BOC, isapproximately 10 kW/m and decreases over the fuel cycle. A largervariation in the differences in LER between the fuel rings was observedpreviously with the reference design,

The relative radial power distributions for the present 62-fuel elementdesign are shown in Table 3. For the present design, the relative powerdensities are nearly the same for the inner and outer ring, within 3% atBOC and within 7% at EOC, and the outer ring produces about 10% moretotal power than the inner ring both at BOC and EOC. The even powerdensity distribution between the inner and outer ring may ensure an evenburnup distribution of the fuel (e.g., exit burnups of approximately 65MWd/kg and 62 MWd/kg, respectively, for the inner and outer fuel ringsat 2.5 m distance from the fuel channel bottom).

TABLE 3 Relative radial power distributions at BOC and EOC. Distancefrom Relative Power Relative Power fuel region Densities Per Ring inlet(m) Inner Ring Outer Ring Inner Ring Outer Ring Beginning of Cycle 0.50.9722 1.0221 0.431 0.569 1.5 0.9818 1.0145 0.435 0.565 2.5 0.99831.0013 0.443 0.557 3.5 1.0062 0.9950 0.446 0.554 4.5 1.0086 0.9932 0.4470.553 End of Cycle 0.5 1.0415 0.967 0.463 0.537 1.5 1.0591 0.953 0.4710.529 2.5 1.0652 0.948 0.473 0.527 3.5 1.068 0.9459 0.475 0.525 4.51.0686 0.9454 0.475 0.525

The flattening of the relative fuel element power densities and burnupprofile in the 62-fuel element design may be considered a significantimprovement over the previous 78-element fuel assembly. The fuelperformance (e.g., thermal conductivity and fission gas retention) maydeteriorate as a function of burnup. An even power and burnupdistribution within the fuel rings therefore may result in improved fuelperformance, as compared to one that favors burnup in the outer fuelring. Local variations in peak fuel and cladding temperatures maycorrelate with variations in radial power distribution. An even powerdistribution between fuel rings which does not change significantly overtime may result in an improved temperature distribution in whichtemperatures are relatively uniform and are unlikely to vary over thecycle.

E. Conclusions

While some changes in the design of the fuel bundle 10 and fuel channelassembly 12 relative to the previous 78-element fuel assembly had anegative impact on the lattice physics (e.g., decrease in reactivity andpositive increase in CVR), a net gain in the lattice physics performanceappears possible. The light water coolant in the central flow passage 20may play a significant role as a moderator. The moderation of neutronsin the central flow passage 20 may result in a significant increase inlattice reactivity and fissile utilization, but also may drive theinfinite lattice and core CVR to be large and negative. The magnitude ofthe CVR may be reduced by reducing the flow area, volume or density ofcoolant in the central flow passage 20. Using core physics modeling,features of the design were found to result in significant improvements,including gains in exit burnup and fissile utilization and reductions inchannel and axial power peaking factors.

2. Thermalhydraulics Analysis

A. Modeling Codes

The peak cladding temperature may be the limiting thermalhydraulicparameter for fuel bundles for the Canadian SCWR, and may be calculatedusing software modeling under different conditions (e.g., modificationsto geometry of the fuel bundle and the fuel channel assembly, beginningof cycle (BOC), and end of cycle (EOC)) based on power distributionsobtained from modeling codes.

In particular, a computer code has been developed to model subchannelflow and phase distribution in a horizontal pressurized heavy waterreactor (PHWR). The code has been designed to be general enough toaccommodate other geometries and orientations. These include singlesubchannels of different shapes, and multiple subchannels of PHWR,pressurized water reactor (PWR) and BWR designs, in both vertical andhorizontal orientations. As well, the code may accommodate a range offluids, including single- and two-phase heavy water, light water,various Freons, and two-phase air-water.

The code has been enhanced to meet the specific requirements for thethermalhydraulic analysis of two-phase flow in the horizontally orientedCANDU (CANada Deuterium Uranium) fuel. The numerical method may modeluni-directional axial flow and bi-directional transverse flow. However,the numerical solution is limited to modeling flow structures in whichthe axial flow is dominant with respect to the lateral flow. Thisprohibits the code from modeling very low axial flow, stagnant flow oraxial flow reversals. This limitation has led to the development of anew staggered grid numerical solution scheme based on apressure-velocity algorithm. This newer version of code has been usedsuccessfully for recirculating flows.

The code version for analysis of the Canadian SCWR includesmodifications used to add three heat transfer correlations and waterproperties for supercritical conditions. This version handles onlysingle-phase calculations, and therefore the transition betweentwo-phase to single-phase or vice-versa is not allowed.

It should be noted that in this analysis the appendages or devices tohold the bundle array and elements are not modeled. The changes to flowdistribution resulting from the inclusion of appendages and grid spacersmay be exploited, for example by enhancing turbulence, and/or divertingthe flow from the inner ring to the middle and outer subchannels, whichmay reduce the subchannel coolant temperature, thus further reducing themaximum wall temperature.

B. Fuel Assembly Description

As described above with reference to FIGS. 1A and 1B, the fuel channelassembly 12 has a re-entrant or double flow pass configuration. Lightwater coolant flows from an inlet plenum into the central flow passage20 located in the inner conduit 14. Bottom ends of the outer conduit 16are sealed, so that when the coolant reaches the bottom of the centralflow passage 20 it is redirected upward and flows through the fuelbundle chamber 18 containing the fuel bundle 10.

The insulator 24 may be arranged between the fuel bundle chamber 18 andthe outer conduit 16, and may be supported on either side by liner tubes26 a, 26 b. If the insulator 24 were to crack, the encapsulation by theliner tubes 26 a, 26 b may ensure that any resultant loose insulatormaterial is not transported by the coolant in the fuel bundle chamber18.

One of the safety features of the Canadian SCWR is the passive removalof long term decay heat through the moderator during postulated loss ofcoolant accidents. Passive decay heat removal occurs through heattransferred to the moderator. Thus, the insulator 24, while minimizingheat transfer to the outer conduit 16 and the moderation region 22during normal operating conditions, may allow sufficient heat transferduring postulated accident conditions to ensure long term decay heatremoval. Design of the insulator 24 may therefore satisfy criteria forpassive long term decay heat removal, with the compromise thatsubcooling of the moderation region 22 may be required during normaloperating conditions.

The Canadian SCWR is intended to operate at 25 MPa, with a coolant inlettemperature of 350° C. and outlet average temperature of 625° C.

Again, exemplary geometric and material specifications for the fuelbundle 10 and the fuel channel assembly 12 are given above in Table 1.

As described herein, in comparison to previous designs, the crosssectional area of the central flow passage 20 is relatively large, forexample, in comparison with the 78-element fuel assembly. This mayresult in a better neutron moderation in the inner conduit 14 caused bythe coolant flowing through the central flow passage 20. The fuel bundle10 has two rings and each one holds thirty one fuel elements 28 a, 28 bper ring, thus having sixty two fuel elements in total. Thisconfiguration may have the following advantages: (i) better neutronmoderation close to the inner conduit 14 which may result in a moreuniform radial power distribution; (ii) discretization of the fuelbundle 10 may be done using only three types of subchannels (spacebetween adjacent fuel elements), namely inner, intermediate and outer;and (iii) a consistent type of subchannels.

C. Comparative Assessment

A comparison with the previous 78-element fuel assembly is presentedherein as part of the assessment of the performance of the 62-elementfuel bundle. Parameters for both bundle designs are presented in Table4.

TABLE 4 Geometrical and thermalhydraulic parameters. 78-element62-element Geometry Number of rings (containing fuel) 3 2 Number ofelements per ring 15/21/42 31/31 Element diameter (mm) 13.6/13.6/8.2 9.5/10.0 Flow tube diameter (mm)  57.6  91.2 Liner internal diameter(mm) 136  144  Fuel volume per bundle (m³) 3.7 × 10⁻² 2.3 × 10⁻² Totalhydraulic diameter (mm)   5.54   7.26 Thermalhydraulic parametersBOC/EOC Power 10103.0/7424.4 9925.7/9274.3 Averaged AFD Peak 1.2308/1.1200 1.1643/1.0038 Maximum Wall Temperature 1265/992835.1/841.6

The comparative analysis is performed using modeling code with theJackson heat transfer correlation for supercritical conditions, theCarlucci mixing model and the flow resistance correlation ofColebrook-White. Comparisons are made under conditions corresponding tothe beginning of cycle (BOC) and end of cycle (EOC). This set of modeloptions is hereafter called the base case.

D. Radial Power Distribution

The relative radial power distributions for the current 62-element fueloption, and the previous 78-element fuel assembly are provided above inTable 3. The even power density distribution between the inner and outerrings may help in reducing hot spots within the fuel bundle 10. This, incombination with uniformity of fuel element diameters and consistentsubchannel sizes, may result in lower wall temperatures.

E. Axial Power Distribution

Referring to FIGS. 9 and 10, the axial power distribution may bedirectly related to the axial location of the maximum wall temperature.For instance, a cosine shape profile may lead to a maximum walltemperature downstream of the peak distribution. A flat powerdistribution may tend to result in a maximum wall temperature at the endof the fuel bundle.

F. Average Outlet Temperature and Maximum Wall Temperature

The wall temperature and coolant temperature may be directly related. Infact, a uniform cross sectional coolant temperature distribution maygive the lowest maximum wall temperature for any arbitrary geometricalarrangement of fuel pins, whereas a non-uniform may result in higherwall temperatures. The following temperature distribution analysis isperformed using the outlet conditions (i.e. at 500 cm).

The cross sectional temperature distributions were determined for boththe fuel bundle 10 and the fuel channel assembly 12 of the presentdisclosure, and the previously investigated 78-element fuel assembly.

The BOC and EOC temperature distributions for the 62-element fuelassembly are shown in FIGS. 11 and 12, respectively. The coolanttemperature ranges from 597° C. to 651° C. for the BOC, and 605° C. to638° C. for the EOC. The inner subchannels are the coldest andcorrespond to a flow tube wall temperature of −600° C. Lowertemperatures may occur in this area because the inner conduit 14 doesnot generate heat. The intermediate and outer subchannel coolanttemperatures are within 20° C. of the average radial temperature. Thecoolant temperature increases in moving from the inner to the outersubchannels.

In contrast to the 62-element fuel assembly, the 78-element fuelassembly has four subchannel rings, the inner, inner intermediate, outerintermediate and outer subchannels. Similar to the 62-element, thelowest temperature is located in the inner ring and increases towardsthe outer ring. However, the coolant temperature ranges from 401° C. to1141° C. for the BOC, and 452° C. to 777° C. for the EOC. Thesignificant difference between the minimum and maximum temperatures maybe due to the non-uniform radial heat flux distribution combined withuneven subchannel sizes (thus resulting in different subchannel massflows).

Based on these results, uniform heat flux distributions may help toreduce hot spots in the subchannels. (This may be achieved with auniform radial power profile and equal fuel element heated perimeter).Furthermore, because the flow tube does not generate heat, the innersubchannels may tend to be the coolest.

G. Sensitivity Analysis

The supercritical heat transfer correlations available in the modelingcode were used to assess their impact on wall temperature predictions.The results are presented in Table 5.

TABLE 5 Sensitivity analysis results. Max wall Axial temp. Sub LocationCase (° C.) Rod # Channel # (cm) Beginning of Cycle Base Case 835.1 3363 430-440 Base Case + 791.9 33 63 440-450 Dittus-Boelter Base Case +867.2 33 63 420-430 Bishop Base Case + 918.1 33 63 410-420 Mokry(Modified Bishop) Base Case + 850.6 33 94 430-440 Offset inner ringangle by 5.6 degrees Base Case 792.3 34 97 430-440 with 32 elements perring End of Cycle Base Case 841.6 33 63 480-490 Base Case + 805.4 33 63480-490 Dittus-Boelter Base Case + 868.9 33 63 480-490 Bishop BaseCase + 907.9 33 63 480-490 Mokry (Modified Bishop) Base Case + 834.9 3394 480-490 Offset inner ring angle by 5.6 degrees Base Case 842.8 2 1480-490 with 32 elements per ring

The results show that the code predicted the lowest wall temperature,using the Dittus-Boelter correlation, which gave temperatures that wereapproximately 30 to 40° C. lower than the base case. The Bishopcorrelation resulted in wall temperatures about 30° C. higher than thebase case. The Mokry correlation resulted in the highest walltemperature predictions, over 80° C. for the BOC and 30° C. for the EOCcompared to the base case.

The sensitivity to element arrangement and number of fuel elements wasalso examined. An angular offset between the inner and outer ring mayresult in a change in the size and shape of both the inner and outersubchannels, which would result in a potential redistribution of theradial heat power distribution. Increasing the number of fuel elementsin each ring may decrease the power per fuel element, and therefore mayreduce the wall temperature. However, due to geometrical constraints,increasing the number of fuel elements may result in a reduction in thetotal fuel mass, thus increasing the power density, potentiallycounteracting the advantage of lower power per fuel element.

To investigate the potential benefits and impacts of varying the offsetof the rings, and/or increasing the number of fuel elements in the innerand outer rings, two more cases were analyzed: a 62-element fuelassembly in which the inner ring of fuel elements was offset by 5.62°;and 64-element fuel assembly with the same offset. The results fromanalysis of these cases are also presented in Table 5 above. Theinclusion of an angular offset of the inner ring in the 62-element fuelassembly resulted in an increase of the maximum wall temperature of 15°C. for the BOC and a decrease of 7° C. for the EOC. In addition, thelocation of the maximum wall temperature shifted from the outer to theintermediate subchannel. The introduction of an additional fuel pin ineach ring, i.e., the 64-element fuel assembly, results in a reduction of40° C. in the peak wall temperature for the for the BOC, and aninsignificant change for the EOC. As was observed in the previous case,the maximum wall temperature shifted to the inner ring. The reasons forthese changes may be new power distributions combined with smallerelement diameters.

H. Conclusions

The flattening of the power distribution profile in the 62-element fuelassembly may result in a significant improvement in performance over theprevious 78-element fuel assembly. The even power distribution withinthe rings of the fuel elements may result in lower wall temperaturescompared to one that has a non-uniform power distribution, because localvariations in peak fuel and cladding temperatures generally correlatewith variations in radial power distribution. Because the even powerdistribution between the rings of the fuel elements may not changesignificantly over time, the improvement in temperature distribution mayalso be unlikely to vary over the cycle.

Subchannel geometry may be important for the 62-element fuel assembly.The reduction to the two rings of the fuel elements 28 a, 28 b (comparedto three, for example) in combination with a relatively large centralmoderating region (the central flow passage 20) may enable a morebalanced heat transfer and coolant mass flow within the subchannels.This improved balance in heat transfer and coolant mass flow, in turn,may enable lower achievable wall temperatures. In addition, the designof the fuel bundle 10 may enable flexibility that may be exploited forfurther improvement. For instance, further modifications to the fuelbundle 10 via an offset of the rings and change in sizes of the fuelelements may enable a further decrease in the maximum wall temperatureand linear power rating, although at the expense of reduced total fuelmass.

While the above description provides examples of one or more processesor apparatuses, it will be appreciated that other processes orapparatuses may be within the scope of the accompanying claims.

1. A fuel assembly for a pressure-tube nuclear reactor, comprising: afuel channel assembly comprising an outer conduit, an inner conduitreceived within the outer conduit and defining an annular fuel bundlechamber therebetween for receiving a flow of a coolant in one direction,the inner conduit comprising a central flow passage for receiving a flowof the coolant in an opposite direction; and a fuel bundle positionedwithin the fuel bundle chamber, the fuel bundle comprising a pluralityof fuel elements, and consisting of an inner ring of the fuel elementssurrounding the inner conduit, and an outer ring of the fuel elementssurrounding the inner ring.
 2. The fuel assembly of claim 1, wherein afirst ratio of a cross sectional area of the coolant in the fuel bundlechamber and the central flow passage to a cross sectional area of thefuel elements is between approximately 2.6 and 7.5.
 3. The fuel assemblyof claim 1, wherein a second ratio of a cross sectional area of thecoolant in the central flow passage to a cross sectional area of thecoolant in the fuel bundle chamber is between approximately 0.8 and 1.3.4. The fuel assembly of claim 1, wherein the inner and outer conduitshave generally circular axial cross sectional shapes.
 5. The fuelassembly of claim 4, wherein the central flow passage is laterallysurrounded by the fuel bundle.
 6. The fuel assembly of claim 5, whereina central axis of the central flow passage is laterally centeredrelative to the fuel bundle.
 7. The fuel assembly of claim 6, whereinthe fuel bundle is rotationally symmetrical about the central axis. 8.The fuel assembly of claim 7, wherein the fuel elements of the innerring are positioned along a first common circumference about the centralaxis, and the fuel elements of the outer ring are positioned along asecond common circumference about the central axis that is concentricwith and laterally outboard of the first common circumference.
 9. Thefuel assembly of claim 8, wherein a number of the fuel elements in theinner ring is equal to a number of the fuel elements in the outer ring.10. The fuel assembly of claim 9, wherein a subchannel distance betweeneach of the fuel elements in the inner ring and the correspondingadjacent one of the fuel elements of the outer ring is approximatelyequal to a subchannel distance between each of the fuel elements in theinner ring.
 11. The fuel assembly of claim 8, wherein the fuel elementshave generally circular axial cross sections.
 12. The fuel assembly ofclaim 8, wherein axial cross sectional areas of each of the fuelelements in the inner ring are different than axial cross sectionalareas of each of the fuel elements in the outer ring.
 13. The fuelassembly of claim 12, wherein the fuel elements of the inner ring have asmaller cross sectional area than the fuel elements of the outer ring.14. The fuel assembly of claim 1, wherein the fuel channel assemblycomprises an insulator that is positioned radially intermediate of thefuel bundle chamber and the outer conduit.
 15. The fuel assembly ofclaim 14, wherein the insulator is encapsulated between inner and outerliner tubes, the outer liner tube being arranged along an interiorsurface of the outer conduit.
 16. The fuel assembly of claim 15, whereinthe insulator is formed of a solid material.
 17. The fuel assembly ofclaim 16, wherein the inner and outer liner tubes are formed ofdifferent materials.
 18. A nuclear reactor comprising a plurality of thefuel assemblies of claim 1 arranged in a lattice, wherein a moderatorregion laterally surrounds the outer conduit of each of the fuelassemblies, the moderator region retaining a moderator therein. 19-25.(canceled)
 26. A fuel assembly for a nuclear reactor, comprising: a fuelchannel assembly comprising an outer conduit, an inner conduit receivedwithin the outer conduit and defining an annular fuel bundle chambertherebetween for receiving a flow of coolant in one direction, the innerconduit comprising a central flow passage for receiving a flow of thecoolant in an opposite direction; and a fuel bundle positioned withinthe fuel bundle chamber, the fuel bundle comprising a plurality of fuelelements, wherein at least one of the following conditions is satisfied:(i) a first ratio of a cross sectional area of the coolant in the fuelbundle chamber and the central flow passage to a cross sectional area ofthe fuel elements is between approximately 2.6 and 7.5; and (ii) asecond ratio of a cross sectional area of the coolant in the centralflow passage to a cross sectional area of the coolant in the fuel bundlechamber is between approximately 0.8 and 1.3.
 27. A nuclear reactor,comprising: a plurality of fuel assemblies arranged in a lattice, eachof the fuel assemblies comprising a fuel channel assembly comprising anouter conduit, an inner conduit received within the outer conduit anddefining an annular fuel bundle chamber therebetween receiving a flow ofa coolant in one direction, the inner conduit comprising a central flowpassage receiving a flow of the coolant in an opposite direction, and afuel bundle positioned within the fuel bundle chamber, the fuel bundlecomprising a plurality of fuel elements; and a moderator regionlaterally surrounding the outer conduit of each of the fuel assemblies,the moderator region retaining a moderator therein, wherein at least oneof the following conditions is satisfied: (i) a first ratio of a crosssectional area of the moderator in the moderator region to a crosssectional area of the fuel elements is between approximately 10 and 20;and (ii) a second ratio of a cross sectional area of the moderator inthe moderator region to a cross sectional area of the coolant in thefuel bundle chamber and the central flow passage is betweenapproximately 2.7 and 3.7. 28-37. (canceled)